• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.717-724
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the VHTR. In this study, structural sizing methodology for the primary HGD with a coaxial double-tube of the VHTR that produces heat at temperatures in the order of $950^{\circ}C$ was suggested and a structural pre-sizing of it was carried out as an example.

• 공업화학
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• 제23권2호
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• pp.204-209
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• 2012

상업용 루테늄 촉매 상에서 에탄의 수증기 개질 반응에 대한 kinetics 데이터를 얻기 위하여 반응온도, 에탄의 분압, 수증기/에탄의 비 등을 변화시키면서 반응 실험을 수행하였다. Kinetics 데이터를 사용하여 Power rate law kinetic model 과 Langmuir-Hinshelwood model의 parameter를 구하였다. 또한 kinetic model식을 적용하여 PRO/II를 이용한 공정 모사를 통해서 에탄의 수증기 개질 반응기 sizing을 수행하였다. 동일한 전환율을 얻기 위해서는 Power rate law model을 적용하였을 경우가 Langmuir-Hinshelwood model을 적용하였을 경우보다 개질 반응기의 부피가 더 큼을 알 수 있었다. Langmuir-Hinshelwood model에 의해 계산된 반응 속도가 반응 실험 결과에 의해 구해진 반응 속도와 더 잘 일치했기 때문에 Langmuir-Hinshelwood model을 적용하여 계산된 반응기의 크기가 실제 반응기 설계에 더 적절하다고 판단된다.

• 청정기술
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• 제19권1호
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• pp.51-58
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• 2013

니켈 촉매 상에서 에탄의 수증기 개질 반응과 수성가스 전환반응 반응에 대한 반응속도 데이터를 얻기 위하여 반응온도와 반응물의 분압을 변화시키면서 반응 실험을 수행하였다. 반응속도 데이터를 사용하여 거듭제곱 속도식 모델(power law kinetic model)과 랭미어-힌쉘우드 모델(Langmuir-Hinshelwood model)의 매개변수를 구하였다. 또한 반응 속도 모델식을 적용하여 PRO/II를 이용한 공정 모사를 통해서 에탄의 수증기 개질 반응기 사이징(sizing)을 수행하였다. 에탄을 반응물로 하여 수증기 개질 반응을 수행한 결과, 단순한 거듭제곱 속도식 모델보다 표면반응에 의하여 반응속도가 결정되는 랭미어-힌쉘우드 모델이 보다 적합하였고, 수성가스 전환반응에 대한 반응속도식은 거듭제곱 속도식 모델이 적합함을 보였다. PRO/II 시뮬레이션을 통해서 수소 생산량에 필요한 반응기의 크기를 결정할 수 있었다.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.558-563
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• 1997

A thermal sizing code, named as HSGSA (Helical coil Steam Generator Sizing Analyzer), for a sodium heated helical coil steam generator is developed for KALIMER (Korea Advanced LIquid MEtal Reactor) design. The theoretical modeling of the shell and tube sides is described and relevant correlations are presented. For assessment of HSGSA, a reference plant design case is compared to the calculational outputs from HSGSA simulation.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.1-6
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2005년도 춘계학술대회 논문집
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• pp.216-227
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• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.95-102
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• 2007

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.